Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes
- Case Western Reserve Univ., Cleveland, OH (United States)
We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.
- Research Organization:
- Case Western Reserve Univ., Cleveland, OH (United States); Univ. of Akron, OH (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE). Nuclear Energy University Programs (NEUP)
- DOE Contract Number:
- AC07-05ID14517
- OSTI ID:
- 1333910
- Report Number(s):
- DOE/NEUP-12-3451; 12-3451; TRN: US1700787
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
AUSTENITIC STEELS
ACCIDENT-TOLERANT NUCLEAR FUELS
FUEL CANS
WATER MODERATED REACTORS
SUPERSATURATION
CARBON
NITROGEN
SOLUTES
WATER COOLED REACTORS
SERVICE LIFE
STAINLESS STEELS
PERFORMANCE
SAFETY
INTERSTITIALS
MECHANICAL PROPERTIES
LOSS OF COOLANT
BWR TYPE REACTORS
NICKEL IONS
PHYSICAL RADIATION EFFECTS
PROTONS
MEV RANGE 01-10
TEMPERATURE RANGE 0400-1000 K
CORROSION RESISTANCE
HYDROGEN EMBRITTLEMENT